The present invention relates to a process for removing technetium from solutions comprising water soluble nuclear waste salts.
Many tens of millions of gallons of high level liquid nuclear waste, in the form of a salt solution containing water-soluble radioactive salts, have accumulated over the past 28 years of operation at the Savannah River Plant, D.O.E., the primary source in the United States for special nuclear materials used in defense, aerospace and medical applications. (Other D.O.E. facilities have similar accumulations.) The radioactivity of the Savannah River salt solution is about 4 Ci/gallon, more than 99% of which is contributed by cesium-137. The remaining radioactivity is from strontium-90, ruthenium-106, plutonium and other isotopes, including technetium-99 (Tc-99) present as the pertechnetate anion (TcO.sub.4.sup.-). The salt solution is stored in carbon steel waste tanks which are unacceptable for permanent storage.
It is desirable to decontaminate the radioactive salt solution and solidify it in a concrete material called "saltstone." A precipitation-adsorption process to decontaminate a salt solution of the sort described above is disclosed in U.S. Pat. No. 4,432,893 (hereinafter "Lee et al"), the contents of which are incorporated herein by reference. In the disclosed process, the major radioactive components of the salt solution, cesium-137, strontium-90 and plutonium, are removed by contacting the radioactive waste solution simultaneously with sufficient sodium tetraphenylborate to precipitate the cesium, and with sufficient sodium titanate to adsorb the strontium and plutonium. The solids can then be separated (for example, by cross-flow filtration) from the decontaminated salt solution, wherein Tc-99 remains as the predominant long-lived radionuclide.
Technetium has a half-life of 2.13.times.10.sup.5 years but does not occur naturally. Technetium-99 is a potential corrosion inhibitor and catalyst for chemical reactions such as hydrocarbon cracking and reforming. However, its use has not been implemented at least in part because Tc-99 is not available in large quantities. The recovery of Tc-99 from aqueous nuclear waste streams is the only source of significant quantities of technetium.
A method for recovering Tc-99 from such waste streams is disclosed by U.S. Pat. No. 3,890,244. In the disclosed method, the acidic waste solution comprising technetium is mixed with an alkaline earth metal oxide or hydroxided flocculant, forming a precipitate and a technetium-containing supernatant phase. After the supernatant is separated from the precipitate and acidified with mineral acid, the technetium is electrolytically removed as an elemental deposit on the cathode of an electrolytic cell.
Technetium removal by the combined use of solvent extraction and electrolysis, as taught in the above-summarized U.S. patent, is a complex process, requiring sophisticated equipment that is difficult to maintain and repair in the context of operating a radioactive system. An alternative method for recovering Tc-99 values from a waste stream is disclosed by Forberg and Westermark in Proc. Int'l. Seminar Chem. & Proc. Eng'rg High-level Liquid Waste Solidification (R. Odvj & E. Merz eds. 1981). The disclosed method entails adding a ferrous salt to a dimethylamine-neutralized, technetium-containing feed solution previously subjected to sorption over a titanate ion exchanger. More specifically, technetium activity in effluent from the titanate ion exchanger is scavenged by the addition of the ferrous salt, followed by precipitation of the hydroxide. This ferrous salt reduction/precipitation approach has the advantage of simplicity, but it cannot be used with the high-pH waste streams (typical pH of about 12 to 14) generated by systems like the Savannah River Plant, apparently because of the lowered reducing potential of ferrous ion at high pH.
Refosco et al, Inorganica Chimica Acta 64:L205 (1982), disclose a method for recovering Tc-99, as [AsPh.sub.4 ]TcO.sub.4, from laboratory wastes by treating aqueous waste solutions with a saturated solution of tetraphenylarsonium chloride; recovery was reported to be nearly quantitative. However, the high cost of the arsonium salt employed by Refosco et al, and the toxicity of the arsenic-containing pertechetate precipitate they obtained, represent substantial drawbacks to adapting the disclosed method to the processing of large amounts of high level liquid nuclear wastes. In addition, the concentration of technetium in the laboratory waste solutions treated by Refosco et al (approximately 7.times.10.sup.-3 M) was substantially higher than that normally encountered in D.O.E. salt solutions (average Tc-99 concentration about 6.times.10.sup.-5 M).